1. Field of the Invention
The present invention relates to zirconium alloys, particularly zirconium alloys for use in nuclear reactor fuel cladding applications, and more particularly zirconium alloy fuel cladding manufactured without requiring conventional late stage heat treatments.
2. Background Art
Nuclear reactors are used in electric power generation, research and propulsion. A reactor pressure vessel contains the reactor coolant, i.e., water, which removes heat from the nuclear core. Piping circuits are used to carry the heated water or steam from the pressure vessel to the steam generators or turbines and to return or supply circulated water or feedwater to the pressure vessel. Typical operating pressures and temperatures for the reactor pressure vessels can be about 7 MPa and 288° C. for BWRs (Boiling Water Reactors) and about 15 MPa and 320° C. for PWRs (Pressurized Water Reactors). The materials used in these respective environments must, in turn, be formulated and/or manufactured to withstand various loading, environmental (high-temperature water, oxidizing species, radicals, etc.) and radiation conditions to which they will be subjected during extended operation of the reactor.
BWR and PWR typically include nuclear fuel sealed in cladding comprising one or more layers of metal or metal alloys to isolate the nuclear fuel from the moderator/coolant system, i.e., water in PWRs and steam and/or water in BWRs. The cladding includes at least one layer of a zirconium-based alloy containing one or more alloying element and can include a second layer of a zirconium alloy. Cladding may also utilize a composite system having an inner lining of sponge zirconium or dilute zirconium alloy containing minor amounts, less than about 0.5 wt % of iron or other elements, as alloying metals. Typically, the cladding will be configured as a tube in which pellets of the nuclear fuel are stacked to fill substantially the entire length of the cladding tube. The tubes will then be arranged in bundles, with a plurality of bundles being arranged to define the reactor core.
Under normal operating conditions, zirconium-based alloys are useful as a nuclear fuel cladding material due to their relatively low neutron absorption cross sections and, at temperatures below about 398° C., their strength, ductility, stability, and lack of reactivity in the presence of demineralized water or steam. “Zircaloys” are a widely used family of commercially-available, corrosion-resistant, zirconium-based alloy cladding materials that include 97-99% by weight zirconium, with the balance being a mixture of some or all of tin, iron, chromium, nickel, oxygen and lesser amounts of carbon, silicon and other unavoidable impurities. Two particular alloy compositions, specifically Zircaloy-2 and Zircaloy-4, are widely used for manufacturing cladding although Zircaloy-2 is the more commonly utilized composition for BWR applications.
In addition to zirconium, Zircaloy-2 includes about 1.2-1.7 wt % Sn; 0.07-0.20 wt % Fe; 0.05-0.15 wt % Cr, and 0.03-0.08 wt % Ni. Zircaloy-4, on the other hand, although including similar quantities of the other alloying elements present in Zircaloy-2, is substantially free of nickel and has an Fe concentration of about 0.18-0.24 wt %.
Tin is useful for improving strength and corrosion resistance of zirconium-base alloys. Zirconium alloys containing less than 0.5 wt. % tin do not tend to provide sufficient strength, while alloys containing more than 2 wt. % tin tend to exhibit decreased corrosion resistance. Nickel is useful for imparting improved corrosion resistance to zirconium alloys. When nickel is incorporated at less than about 0.03 wt. %, the improvement in corrosion resistance is marginal, while alloys containing more than 0.2 wt. % tend to exhibit degraded hydrogen absorption characteristics. Like nickel, chromium is useful for imparting improved corrosion resistance to zirconium alloys. The quantity of chromium incorporated in the alloy affects the relative proportions of Zr2(Fe, Ni) and Zr(Fe, Cr)2 type precipitates that, in turn, affects the corrosion resistance and hydrogen absorption properties of the alloy. Further, although iron can be used to increase corrosion resistance, alloys including more than about 0.6 wt % iron tend to exhibit degraded working characteristics, particularly ductility, so the utility of iron in this regard is limited.
The presence of these alloying elements, which are relatively insoluble in zirconium under normal conditions, will generally result in the formation of intermetallic Second Phase Particle (SPP) “precipitates” in an α-phase zirconium matrix when alloying elements are present in concentrations above their solubility limits. For example, the precipitates most commonly found in Zircaloys may be generally represented by the chemical formulas Zr(Fe,Cr)2 and Zr2(Fe,Ni).
Coolant, typically demineralized water that may include a variety of treatment compositions or packages, will generally flow through flow channels provided between and/or along the fuel elements to remove heat from the core. The cladding, by separating the nuclear fuel from the coolant, prevents or reduces the amount of radioactive fuel and fission products from entering the coolant stream and being spread throughout the primary cooling system. Degradation of the integrity of the cladding layer, whether by cracking and/or corrosion, may result in containment failures that allow contamination of the coolant.
In addition to the basic composition of the alloy, conventional techniques for fabricating fuel cladding include solution heat treatment methods in which an alloy is heated, for a short period of time, to a temperature at which the alloy exists in α+β or β phase and the intermetallic particles become dissolved, after which the alloy is rapidly quenched, at which point the intermetallic particles can be re-nucleated and grown. Such a process is described in Japanese Patent Publications Nos. 45699/1986, 58223/1988 and 3172731/2001. Other heat treatment methods have been described in U.S. Pat. Nos. 4,450,016, 4,576,654, and 5,437,747, each of which is incorporated by reference herein in its entirety. These patents have focused on the control of process parameters, specifically, the heat treatment conditions with the primary intent of controlling the corrosion resistance through microstructural factors. The major deficiency of these patents is in not recognizing the importance of restrictive alloy chemistry control. Furthermore, in the case of 3172731/2001, the emphasis is on controlling the post-solution heat treat thermal exposure through the use of the “Sigma-A” or “ΣA” parameter, which is a measure of the cumulative thermal exposure during fabrication as defined by the formula (I):A=Σtiexp(Q/RTi)  (I)where ti and Ti are the time and temperature of the ith thermal exposure (where t is time in hours; T is temperature in K; Q is the activation energy; R is the gas constant; and Q/R=40,000 K).
However, the use of this parameter solely to control SPP size may not be effective, as differences in process parameters, such as the solution treatment temperature and quench rate, could affect the SPP size for given Sigma-A value or post-solution heat treatment conditions. The application of heat treatment in connection with a particular alloy composition has been detailed in Japanese Laid-Open Patent Publication No. 228442/1987. In this case, the exceptionally high iron and nickel concentrations allowed contribute to Zr2(Fe, Ni) SPP sizes between 100 and 500 nm and the formation of an unusual Sn—Ni compound.
While efforts to improve claddings have continued, the performance of claddings, particularly with regard to corrosion and hydrogen pickup as a result of interactions between the cladding, the nuclear fuel, the radiation field and the coolant continues to be a concern and a maintenance issue in BWRs. One type of continuing corrosion problem associated with BWR operation is nodular corrosion wherein localized regions on the alloy surfaces form lenticular corrosion layers. This nodular corrosion initiates relatively quickly upon operation in a BWR and proceeds under and may be exacerbated by the continuing irradiation of the alloy surfaces. Nodular corrosion can generally be controlled, particularly in aggressive water chemistry environments, as disclosed in a separate application bearing application Ser. No. 10/935,157 and filed on even date herewith, by an astute combination of alloy composition control, fabrication parameters leading to an intermetallic particle size less than about 40 nm, and smooth surface finish.
One problem however with conventional fuel cladding formulations using intermetallic particle sizes below about 40 nm is the need to provide a late stage solution heat treatment and rapid quench process to achieve the desired result. Such a process is complicated and it requires special equipment, process steps, process controls and quality inspection tasks. The use of such a process is costly.
To overcome this inefficiency it is desirable to fabricate fuel cladding without a late stage solution heat treatment. This would lead to a relative increase in the mean intermetallic particle size when compared to a fuel cladding fabricated identically except for the late stage heat-treatment. For an exemplary process defined by the current invention, the mean intermetallic particle size would be no less than about 50 nm and no more than about 100 nm, limited in part by particle coarsening (Ostwald ripening) that is inherent in conventional cladding fabrication processes that include hot extrusion and multiple stages of cold reduction and thermal recrystallization annealing.
The increase in the mean intermetallic particle size would in turn increase the fuel cladding susceptibility to nodular corrosion, even in typical non-aggressive water chemistry environments.
To help overcome this negative aspect of increased nodular corrosion susceptibility associated with non-heat treated cladding and the associated increase in intermetallic particle size, the maximum allowable mean intermetallic particle size must be limited. While developing such a Zircaloy-2 fuel cladding, this upper limit was estimated to be about 100 nm. Arising from these developmental activities for such a Zircaloy-2 fuel cladding, the importance of specific alloy compositions in maximizing nodular corrosion resistance was discovered. Although the importance of alloy composition in providing corrosion resistance is generally known to those skilled in the art, the specific thresholds required for non-heat treated fuel cladding were not known explicitly, nor could they be predicted explicitly.
Besides the inefficiency of fabrication processes utilizing late stage heat treatments, another problem for heat treated cladding related to corrosion is the evolution of the intermetallic particles under the effects of the neutron radiation field present in a nuclear reactor. In this environment, neutrons impinge on the intermetallic particles and cause them to dissolve and/or amorphize. With a simplified view, the extent of this process is dependent on the neutron energy, the cumulative neutron flux (fluence), and the initial particle size and structure. Components with a relatively small mean SPP size of about 25 nm may dissolve completely within a fast neutron fluence of 8.5×1025 n/m2 (E>1 MeV) in a BWR, which is less than half as long as the currently desired lifetime for Zr-alloy components. Inasmuch as the intermetallic particles affect corrosion resistance, their evolution and disappearance can affect the corrosion resistance in a negative manner.
Furthermore, inasmuch as component hydrogen generation and absorption (hydriding) follows directly from the corrosion process, SPP evolution can also affect component hydriding in a negative manner. Although such a negative effect of microstructural evolution on both corrosion and hydriding has been documented in the field, the current Applicants' experience with small-particle fuel cladding corrosion in a BWR has been contrary to the conventional understanding promoted by data collected and widely disseminated by F. Garzarolli and others of particular influence in the field as reflected in, for example, Garzarolli, F. Schumann, R., and Steinberg, E., “Corrosion Optimized Zircaloy for Boiling Water Reactor (BWR) Fuel Elements,” Zirconium in the Nuclear Industry: Tenth International Symposium, ASTM STP 1245, A. M. Garde and E. R. Bradley, Eds., American Society for Testing and Materials, Philadelphia, 1994, pp. 709-23, the contents of which are hereby incorporated by reference in their entirety.
Because heat treated cladding would tend to have smaller particle sizes than non-heat treated fuel cladding, heat treated fuel cladding would experience relatively quicker particle dissolution than non-heat treated fuel cladding. Thus non-heat treated fuel cladding of this invention, although being generally more susceptible to nodular corrosion, may have some benefit in delaying microstructural evolution-induced changes in corrosion resistance and hydriding that may occur at the later exposure stages in a component's operational life.